• Nuclear Engineering and Technology
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• 제41권8호
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• 한국안전학회지
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• 제35권5호
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• pp.121-127
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• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1291-1302
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• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Ocean and Polar Research
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• 제43권3호
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• pp.141-148
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• 2021

Microplastic contamination in waterbodies is a growing source of concern for researchers and other stakeholders. We investigated oxidative stress and toxicity in goldfish (Carassius auratus) in response to exposure to 1-㎛ diameter micro-polystyrene (MP) at concentrations of 0, 10, 100, and 1000 beads/mL (MP 0, MP 10, MP 100, and MP 1000 groups) for 7 d (at day 0, 1, 3, 5, and 7). We analyzed the survival rates; superoxide dismutase (SOD) and catalase (CAT) mRNA expression levels in the liver; SOD and CAT activity in the plasma; caspase-3 mRNA expression in the liver; and the levels of hydrogen peroxide (H₂O₂) in plasma. Terminal transferase dUTP nick end labeling (TUNEL) assays were also conducted to determine apoptosis levels in the liver. All fish in the MP 1000 group died by day 7 and the MP 100 group had a lower survival rate than the MP 10 and MP 0 groups. The mRNA expression as well as SOD, CAT, and caspase-3 activity levels were increased significantly with increases in MP concentration and exposure time. Finally, according to the TUNEL assay, more apoptosis was observed in the MP 1000 group at day 5 than in other groups. In summary, MP concentrations above 100 beads/mL caused death and oxidative stress to goldfish. We conclude that MP can cause oxidative stress and apoptosis in goldfish, which leads to death.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• 한국안전학회지
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• 제29권3호
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• pp.129-135
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• 2014

The number of abnormal operation procedures (AOPs) for mitigating a plant abnormal status amounts to about one hundreds for the most of 1000MWe optimized power plant (OPR1000) and it is expected that the number of AOPs would be increased to cope with an abnormal status occurred newly. However, it is not well organized for operators to select a proper AOP from alarms occurred in main control room (MCR) during a plant abnormal status. It may be a burden to operators since the selection of AOP to respond an abnormal status is authorized by operators. When multiple alarms occur in MCR, it would take more time to respond them than a single alarm. To reduce the efforts, various MCR operation support systems have been developed. The purpose of this study was to develop a multi-alarm pattern card to select an appropriate AOP effectively when multiple alarms occurs in a single upper layout (UL) of MCR. It can be applied for an operation support tool as well as an education tool.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4607-4616
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• 2023

We implement machine learning regression models to predict peak pressures of primary and secondary systems, a major safety concern in Loss Of Condenser Vacuum (LOCV) accident. We selected the Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code to analyze the LOCV accident, and the reference plant is the Korean Optimized Power Reactor 1000MWe (OPR1000). eXtreme Gradient Boosting (XGBoost) is selected as a machine learning tool. The MARS-KS code is used to generate LOCV accident data and the data is applied to train the machine learning model. Hyperparameter optimization is performed using a simulated annealing. The randomly generated combination of initial conditions within the operating range is put into the input of the XGBoost model to predict the peak pressure. These initial conditions that cause peak pressure with MARS-KS generate the results. After such a process, the error between the predicted value and the code output is calculated. Uncertainty about the machine learning model is also calculated to verify the model accuracy. The machine learning model presented in this paper successfully identifies a combination of initial conditions that produce a more conservative peak pressure than the values calculated with existing methodologies.

• Ocean and Polar Research
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• 제37권3호
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• pp.201-209
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• 2015

To examine the effects of increased $CO_2$ concentration and seawater temperature on the photosynthesis and growth of forest forming Ecklonia cava (Laminariales, Phaeophyta), sporophytic discs and gametophytes were cultured under three $pCO_2$ concentrations (380, 750, 1000 ppm), four temperatures (5, 10, 15, $20^{\circ}C$ for sporophytes; 10, 15, 20, $25^{\circ}C$ for gametophytes), and two irradiance levels (40, $80{\mu}mol$ photons $m^{-2}s^{-1}$) for 5 days. Photosynthetic parameter values ($ETR_{max}$, $E_k$, and ${\alpha}$) were generally higher as sporophytic discs were grown under low temperature and increased $CO_2$ concentration at 750 ppm. However, photosynthesis of Ecklonia sporophytes was severely inhibited under a combination of high temperature ($20^{\circ}C$) and 1000 ppm $CO_2$ concentration at the two photon irradiance levels. The growth of gametophytes was maximal at the combination of 380 ppm (present seawater $CO_2$ concentration) and $25^{\circ}C$. Minimal growth of gametophytes occurred at enriched $pCO_2$ concentration levels (750, 1000 ppm) and high temperature of $25^{\circ}C$. The present results imply that climate change which is increasing seawater temperature and $pCO_2$ concentration might diminish Ecklonia cava kelp beds because of a reduction in recruitments caused by the growth inhibition of gametophytes at high $pCO_2$ concentration. In addition, the effects of increased temperature and $pCO_2$ concentration were different between generations - revealing an enhancement in the photosynthesis of sporophytes and a reduction in the growth of gametophytes.